ML20204E494

Discusses TSs Bases Change Re 3/4.4.1,3/4.6.1.7,3/4.6.3, 3/4.7.12 & 3/4.8.4.Forwards Affected Bases Pp B 3/4 4-1, B 3/4 6-3,B 3/4 6-4,B 3/4 7-7 & B 3/4 8-3

ML20204E494
Person / Time
Site:	Waterford
Issue date:	03/17/1999
From:	Chandu Patel NRC (Affiliation Not Assigned)
To:	Dugger C ENTERGY OPERATIONS, INC.
References
TAC-MA0676, TAC-MA676, NUDOCS 9903250046
Download: ML20204E494 (8)

Text

- .,

c Mr. Charles M. Dugger March 17, 1999 Vice President Operations Entergy Operations, Inc.

P. O. Box B -

Killona, LA 70066

SUBJECT:

TECHNICAL SPECIFICATIONS BASES CHANGE FOR WATERFORD STEAM ELECTRIC STATION, UNIT 3 (TAC NO. MA0676)

Dear Mr. Dugger:

By letter dated November 13,1997, Entergy Operations, Inc. (EOl), submitted several changes to the Waterford Steam Electric Station, Unit 3, Technical Specifications (TS) and Bases.

These changes include TS Bases Section 3/4.4.1, " Reactor Coolant Loops and Coolant Circulation"; 3/4.6.1.7, " Containment Ventilation System"; 3/4.6.3, " Containment Isolation

. Valves"; 3/4.7.12, " Essential Services Chilled Water System"; and 3/4.8.4, " Electrical Equipment Protective Devices." These Bases changes were caused by the design changes,

' inconsistencies between the TS and Bases, the addition of information for clarification, and a change in a reference document. EOl evaluated these changes pursuant to 10 CFR 50.59 and determined that these changes do not involve an unreviewed safety question. The staff has no objection to these Bases changes. Enclosed are the affected Bases pages, B 3/4 4-1, B 3/4 6 3, B 3/4 6-4, B 3/4 7-7, and B 3/4 8-3.

The TS changes requested by your November 13,1997, letter are being addressed under separate correspondence.

Since' rely, "0RIGINAL SIGNED BY:

Chandu P. Patel, Project Manager Project Directorate IV-1 Division of Licensing Project Management  !

Office of Nuclear Reactor Regulation Docket No. 50-382 DISTRIBUTION i Docket Filei OGC l

Enclosure:

As stated PUBLIC ACRS l PDIV-1 RF G. Hill (2) cc w/ encl: See next page G. Dick K.Brockman,RIV -

C.Patel J.Zwolinski/S. Black LBerry W.Beckner Document Name: G:\WATFINAL\LTRA0676.WPD '

OFC LA/PDIV-1e - (A)D/PDly-1 I PM/PDIV-1 /

NAME C.PA 4 L. BERRY k I

G. DICK I

'DATE 3 / ll /98 Ob/ /98 3 / // /98 COPY khO [ESlNO YES/NO

~ OFFICIAL R8 CORD COPY .

t r m p _, g ,_.

cg,,

9903250046 990317

_n " {

"-- ---- --% u,., / y PDR ADOCK 05000382 P PDR b

~

urg .

'p  % UNITED STATES

5 P; NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30506-0001

'+,***** March 17, 1999 Mr. Charles M. Dugger Vice President Operations Entergy Operations, Inc. j P. O. Box B Killona, LA 70066 -

SUBJECT:

TECHNICAL SPECIFICATIONS BASES CHANGE FOR WATERFORD STEAM ELECTRIC STATION, UNIT 3 (TAC NO. MA0676)

Dear Mr. Dugger:

By letter dated November 13,1997, Entergy Operations, Inc. (EOI), submitted several changes to the Waterford Steam Electric Station, Unit 3, Technical Specifications (TS) and Bases.

These changes include TS Bases Section 3/4.4.1," Reactor Coolant Loops and Coolant Circulation"; 3/4.6.1.7, " Containment Ventilation System"; 3/4.6.3, " Containment Isolation Valves"; 3/4.7.12, " Essential Services Chilled Water System"; and 3/4.8.4, " Electrical Equipment Protective Devices." These Bases changes were caused by the design changes, inconsistencies between the TS and Bases, the addition of information for clarification, and a {

change in a reference document. EOl evaluated these changes pursuant to 10 CFR 50.59 and  !

determined that these changes do not involve an unreviewed safety question. The staff has no objection to these Bases changes. Enclosed are the affected Bases pages, B 3/4 4-1, B 3/4 6-3, B 3/4 6-4, B 3/4 7-7, and B 3/4 8-3.

The TS changes requested by your November 13,1997, letter are being addressed under separate correspondence.

Sincerely, hhh Chandu P. Patel, Project Manager  ;

Project Directorate IV-1 i Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosure:

As stated cc w/ encl: See next page

Mr. Charles M. Dugger Entergy Operations, Inc. Waterford 3 l

cc: I Administrator Regional Administrator, Region IV Louisiana Radiation Protection Division U.S. Nuclear Regulatory Commission Post Office Box 82135 611 Ryan Plaza Drive, Suite 1000 Baton Rouge, LA 70884-2135 Arlington,TX 76011 Vice President, Operations Resident inspector /Waterford NPS Support .

Post Office Box 822 i Entergy Operations, Inc. Killona, LA 70066  ;

P. O. Box 31995 ~

Jackson, MS 39286 Parish President Council St. Charles Parish Director P. O. Box 302 Nuclear Safety & Regulatory Affairs Hahnville, LA 70057 Entergy Operations, Inc.

P. O. Box B Executive Vice-President Killona, LA 70066 and Chief Operating Officer Entergy Operations, Inc.

Wise, Carter, Child & Caraway P. O. Box 31995 P. O. Box 651 Jackson, MS 39286-1995 Jackson, MS 39205 Chairman General Manager Plant Operations Louisiana Public Service Commission Entergy Operations, Inc. One American Place, Suite 1630 P. O. Box B Baton Rouge, LA 70825-1697 Killona, LA 70066 Licensing Manager Entergy Operations, Inc.

P. O. Box B Killona, LA 70066 Winston & Strawn 1400 L Street, N.W.

Washington, DC 20005-3502

E . .

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor  !

coolant pumps in operation, and maintain DNBR above 1.20 during all normal operations and I anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this i specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE. ,

in MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops or trains (either shutdown cooling or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling trains be OPERABLE.

The operation of one reactor coolant pump or one shutdown cooling (Iow pressure safety injection) pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a reactor coolant pump in MODES 4 and 5, with one or more RCS cold legs less than or equal to 272*F are provided to prevent RCS pressure transients, l caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the ]

. limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing I a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100*F above each of the RCS cold leg temperatures.

3/4.4.2 SAFET(VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 4.6 x 105 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate 6 relieve any overpressure condition which could occur during shutdown. In the event that no safety WATERFORD - UNIT 3 B 3/4 4-1 Revised by NRC letter dated l l March 17, 1999

CONTAINMENT SYSTEMS BASES 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM (Continued) l Leakage integrity testn with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the  !

opportunity for repair before gross leakage failure develops. The 0.60 L. leakage limit shall not be I exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

Operability concems for purge supply and exhaust isolation valves other than those addressed in Actions 'a' and "b" of Specification 3.6.1.7 are addressed under Specification 3.6.3,

' Containment isolation Valves."

d

)

3/_4 ' ' DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT SPRAY SYSTEM and CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Spray System and the Containment Cooling System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or MSLB for any double-ended break of the largest reactor coolant pipe or main steam line.

Under post-accident conditions these systems will maintain the containment pressure below 44 psig and temperatures below 269.3F during LOCA conditions or 413.5'F during MSLB conditions. The systems also reduce the containment pressure by a factor of 2 from its post-accident peak within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, resulting in lower containment leakage rates and lower offsite dose rates.

The Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere under post-LOCA conditions to maintain doses in accordance with 10 CFR 1 Part 100 limits as described in Section 6.5.2 of the FSAR. l In MODE 4 when shutdown cooling is placed in operation, the Containment Spray System is l realigned in order to allow isolation of the spray headers. This is necessary to avoid a single failure of I the spray header isolation valve causing Reactor Coolant System depressurization and inadvertent l spraying of the containment. To allow for this realignment, the Containment Spray System may be  !

taken out-of-service when RCS pressure is s 400 psia. At this reduced RCS pressure and the reduced temperature associated with entry into MODE 4, the probability and consequences of a LOCA or MSLB are greatly reduced. The Containment Cooling System is required OPERABLE in MODE 4 and is available to provide Depressurization and cooling capability.

A train of Containment Cooling consists of two fans (powered from the same safety bus) and their associated coolers (supplied from the same cooling water loop.) One Containment Cooling train and Containment Spray train has sufficient capacity to meet post accident heat removal requirements. j Operating each containment cooling train fan unit for 15 minutes and verifying a cooling water flow rate of 625 gpm ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected and corrective action taken.

WATERFORD- UNIT 3 B 3/4 6-3 Revised by NRC letter dated l March 17,1999

CONTAINMENT SYSTEMS BASES 3/4.6.2.1 and 3/4 6.2.2 CONTAINMENT SPRAY SYSTEM and CONTAINMENT COOLING SYSTEM l (cont'd) j l

The 18 month Surveillance Requirement verifies that each containment cooling fan actuates I upon receipt of an actual or simulated SIAS actuation signal. The 18 month frequency is based on engineering judgment and has been shown to be acceptable through operating experience.

Verifying a cooling water flow rate of 1200 gpm to each cooiing unit provides assurance that the design flow rate assumed in the safety analyses will be achieved. The safety analyses assumed a cooling water flow rate of 1100 gpm. The 1200 gpm requirement accounts for measurement instrument uncertainties and potential flow degradation. Also considered in selecting the 18 month frequency were the know reliability of the Cooling Water System, the two train redundancy, and the low probability of a significant degradation of flow occurring between surveillances. The flow measurement for the 18 month test shall be done in a configuration equivalent to the accident lineup to ensure that in an accident situation adequate flow will be provided to the containment fan coolers for them to perform their safety function.

Verifying that each valve actuates to the full open position provides further assurance that the valves will travel to their full open position on a Safety injection Actuation Signal.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through GDC 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

The opening of locked or sealed closed containment isolation valves on a intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

" Containment Isolation Valves", previously Table 3.6-2, have been incorporated into the Technical Requirements Manual (TRM).

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the WATERFORD- UNIT 3 B 3/4 6 4 Revised by NRC letter dated l March 17,1999

1 PLANT SYSTEMS BASES 3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probabks of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shield mechanism.

3/4.7.10 THIS SECTION DELETED 3/4.7.11 THIS SECTION DELETED 3/4.7.12 ESSENTIAL SERVICES CHILLED WATER SYSTEM The OPERABILITY of the essential services chilled water system ensures that sufficient chilled water is supplied to those air handling systems which cool spaces containing equipment required for safety-related operations.

l WATERFORD- UNIT 3 B 3/4 7-7 gsg gterdatedl

ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

Operation with a battery ceirs parameter outside the normallimit but within the allowable value specified in Table 4.8 2 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability., (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturers recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual celis specific gravity, ensures that an individual ceirs specific gravity will not be more than 0.040 below the manufacturers full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit., and (4) the allowable value for an individual cetrs float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance.

The Surveillance Requirements applicable to lower voltage circuit breakers and fuses provides assurance of breaker and fuse reliability by testing at least one representative sample of each manufacturers brand of circuit breaker and/or fuse. Each manufacturers molded case and metal case circuit breakers and/or fuses are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers and/or fuses are tested. If a wide variety exists within any manufacturefs brand of circuit breakers and/or fuses it is necessary to divide that manufacturers breakers and/or fuses into groups and treat each group as a separate type of breaker or fuses for

, surveillance purposes.

The OPERABILITY of the motor-operated valves thermal e ed protection and/or bypass devices ensures that these devices will not prevent safety related ' % o from performing their function.

The Surveillance Requirements for demonstrating the OPERABILI. cf these devices are in accordance with Regulatory Guide 1.106 " Thermal Overload Protection for Electric I Aotors on Motor Operated Valves," Revision 1, March 1977.

" Containment Penetration Conductor Overcurrent Protection Devices" and " Motor-Operated Valves Thermal Overload Protection and/or Bypass Devices", prev'ously Tables 3.8-1 and 3.8 2, have been incorporated into the Technical Requirements Manual (TRM) l l WATERFORD- UNIT 3 B3/48-3 g p b g tter datal l

- -- - .-